ML20116J307
| ML20116J307 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 08/07/1996 |
| From: | Kelly G NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20116J308 | List: |
| References | |
| NUDOCS 9608130142 | |
| Download: ML20116J307 (14) | |
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UNITED STATES
.40 CLEAR REGULATORY COMMISSION 2
WASHINoTON, D.C. 206WW201
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I IES UTILITIES INC.
CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET No. 50-331 DUANE ARNDLD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICDISE Amendment No. 215 License No. DPR-49 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by IES Utilities Inc., et al.,
dated November 30, 1995, complies with the standards and l
requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; 8.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conductied without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulati3ns; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-49 is hereby amended to read as follows:
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9608130142 960807 PDR ADOCK 05000331 P
. (2) Technical Snecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 215. are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective as of the date of issuance and shall be implemented within 120 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMISSION
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Glenn B. Nelly, roj t Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance:
August 7, 1996
ATTACMENT TO LICENSE AMENDMENT NO. pic; FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by vertical lines.
Remove Insert vil vil 1.1-11 1.1-11 1.1-12 1.1-12 3.3-6 3.3-6 3.3-7 3.3-7 3.3-7a 3.3-7a 3.3-7b 3.3-7b 3.3-13 3.3-13 3.3-13a 3.3-15 3.3-15 3.3-16 delete 6.11-4 6.11-4 I
W DAEC-1 TECHNICAL SPECIFICATIONS LIST OF FIGURES FIGURE NUMBEP.
TITLE 1.1-1 Power / Flow Map 2.1-1 APRM Flow Biased Scram and Rod Blocks 4.1-1 Instrument Test Interval Determination Curves 4.2-2 Probability of System Unavailability Vs. Test Interval 3.4-1 Sodium Pentaborate Solution Volume Concentration Requirements 3.4-2 Minimum Temperature of Sodium Pentaborate Solution 3.6-1 DAEC Operating Limits 4.8.E-1 DAEC Emergency Service Water Flow Requirement i
Amendment No. 451,154,100,107, vil
+90183-MM 215
DAEC-1 iv.
The analytical procedures now used result in more logical answer than the alternative method of assuming a higher starting power in conjunction with the expected values for l
the parameters.
Trio Settinas The bases for individual trip settings are discussed in the following paragraphs.
A.
Neutron Flux Trips 1.
APRM Hioh Flux Scram (Run Mode)
The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power (1658 MWt).
Because fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel. Therefore, during abnormal operational transients, the thermal power of the fuel will be less than the indicated by the neutron flux at the scram setting. Analyses are performed to demonstrate that the APRM flux scram over the range of settings from a maximum of 120E to the minimum flow biased setpoint of 62% provide protection from the fuel safety limit for all abnormal operational transients including those that may result in a thermal hydraulic instability.
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Amendment No. 44h 215 1.1-11 f
DAEC-1 An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is reached. The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation.
Reducing this operating margin would increase the frequency of spurious scrans which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the APRM scram trip setting was selected because it provides adequate margin for the fuel cladding integrity Safety Limit yet allows operating margin that reduces the possibility of unnecessary scrams.
Amendment No. He,215 1.1-12
DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.
If Specification 3.3.D.1, 2 or 3 cannot be met, be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
^
E.
Reactivity Anomalies E.
Reactivity Anomalies The reactivity difference The rod density shall be between the actual rod density predicted and compared to the and predicted rod density actual rod density:
shall not exceed 1% Ak/k.
1.
If the reactivity is different 1.
During the first startup by more than 1% Ak/k, Perform following CORE ALTERATIONS and an analysis to determine and explain the cause of the reactivity difference; operation may continue if the difference is explained and corrected.
2.
Otherwise be in COLD SHUTDOWN 2.
At least once per full power within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
month.
F.
Recirculation Pumos F.
Recirculation Pumps 1.
Operation in natural 1.
Not used circulation
- is not permitted.
I If operation in natural circulation occurs, the
{
reactor shall be scrammed.
1 2.
No recirculation pump shall be 2.
Not used placed in operation while the i
reactor is in natural circulation.
3.
Operation in the Exclusion 3.
Not used Region of the power / flow map described in the Core Operating Limits Report is not permitted.
If entry into this region occurs, immediately insert control rods or increase core flow to exit the region.
4
- No recirculation pumps running and two or more control rods withdrawn and the reactor in STARTUP or RUN.
Amendment No. H4,117,141,100, 3.3-6
-183, 215
DAEC-1 LIMITING COM)ITIONS FOR OPERATION ElRYJILLANCE REOUIREMENTS I
4.
Single Loop Operation (SLO) 4.
Single Loop Operation (SLO)
The reactor may be started and operated, or may continue operating in SLO provided the following restrictions are observed:
a.
NAPLNGR multipliers and MCPR a.
Jet Pump baseline data for SLO adjustment are used in shall be updated as soon as accordance with the CORE practical after entering SLO OPERATING LIMITS REPORT.
per Specification 4.6.E.4.
b.
Flow Biased APRM setpoints are adjusted for SLO per Specifications 3.1.A and 3.2.C.
i c.
The idle loop is isolated electrically by disconnecting the breaker to the recirculation pump motor generator (M/G) set drive motor prior to reactor startup, or if disabled during reactor operation, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of entering SLO.**
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4
- The breaker may be racked in and the M/G set and recire. pump started under administrative control for testing provided Specification 3.3.F.5 is satisfied.
i Amendment No. 414,117,141,180, 3.3-7 463r-215 s
DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0VIREMENTS 1
I 5.
Restoration from SLO a.
Verify the thermal limitations of Specification 3.6.A are met ~
prior to startup of the idle recirculation loop.
b.
After startup of the idle recirculation pump, the discharge valve of the lower speed pump may not be opened unless the speed of the faster pump is less than 50% of its rated speed.
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4 Amendment No. 144,117,141,48G, 3.3-7a
-183; 215
I DAEC-1 (DELETED) i I
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1 Amendment No. +14d4hl41,180,18h 3.3-7b 215
1 DAEC-1 The REM bypass time delay is set low enough to assure minimum rod j
movement while upscale trips are bypassed.
I A Limiting Control Rod Pattern for rod withdrawal error (RWE) exists when (a) core thermal power is greater than or equal to 30%
4 of rated and less than 90% of rated (30% $ P s 90%) and the MCPR i
is less than 1.70, or (b) core thermal power is greater than or j
equal to 90% of rated (P ;t 90%) and the EPR is less than 1.40.
l During the use of such patterns, it is judged that testing of the REM channel (when one channel is inoperable) prior to withdrawal of such rods to assure its operability will assure that improper withdrawal does not occur.
D.
Scram Insertion Times The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent fuel damage; i.e., to prevent the MCPR from becoming less than the safety limit.
i; After initial fuel loading and subsequent refuelings when operating above 950 psig, all control rods shall be scram tested within the constraints imposed by the Technical Specifications and before the 40%
power level is reached. The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.
i E.
Reactivity Anomalies During each fuel cycle excess operative reactivity varies as fuel-depletes and as any burnable poison in supplementary control is burned.
The magnitude of this excess reactivity may be inferred from the
)
critical rod configuration. As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected by comparison of the critical rod pattern at selected base states to the predicted rod inventory at that state.
Power operating base conditions provide the most sensitive and directly interpretable data relative to core reactivity.
Furthermore, using power operating base conditions permits frequent reactivity comparisons.
Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds i
1% ak/k. Deviations in core reactivity greater than 1% Ak/k are not expected and require thorough evaluation. One percent reactivity limit is considered safe since an insertion of the reactivity into the core would not lead to transients exceeding design conditions of the reactor system.
F.
Recirculation Pumps Not allowing startup of an idle recirculation punp from a natural circulation condition prevents the reactivity insertion transient that would occur due to the sudden flow of cold stratified water into the core.
In addition, operation in natural circulation could place Amendment No. 114,110,120,141, 3.3-13 142,100,103,- 215
DAEC-1 the plant in or near the exclusion region.
Restarting a recirculation pump while in the exclusion region could result in the initiation of i
thermal hydraulic instability. Manually scramming the reactor is the l
j recommended method of exiting the exclusion region when the plant is i
operating in natural circulation.
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The reactor design criteria is such that thermal hydraulic oscillations I
are prevented or can be readily detected and suppressed without l
exceeding specified fuel design limits. To minimize the likelihood of l
an instability, a power / flow exclusion region to be avoided during normal operation is calculated using the approved methodology as stated in Specification 6.ll.2.a.5.
Since the exclusion region may change each fuel cycle, the limits are contained in the Core Operating Limits Report.
Specific directions are provided to avoid operation in this i
region and to immediately exit upon an entry.
Entries into the exclusion region are not part of normal operation. Any entry may occur i
as a result of an abnormal event, such as a single recirculation pump trip.
In these events, operation in the exclusion region may be needed l
to prevent equipment damage, but actual time spent inside the exclusion region is minimized. Though each operator action can prevent the occurrence and protect the reactor from an instability, the APRM flow-i biased scram function is designed to suppress global oscillations, the most likely mode of oscillation, prior to exceeding the fuel safety limit. While global oscillations are the most likely mode, protection i
from out-of-phase oscillations are provided through avoidance of the l
exclusion region and administrative controls on reactor conditions which j
are primary factors affecting reactor stability.
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i Amendment No. 215 3.3-13a
DAEC-1 3.3 and
4.3 REFERENCES
e 1.
Banked Position Withdrawal Seouence, NED0-21231, January 1977.
2.
General Electric Standard Anolication for Reactor Fuel, NEDE-24011-P-A*.
3.
General Electric Service Information Letter (SIL) No. 316, Reduced Notch Worth Procedure, November 1979.
4.
Averaae Power Ranae Monitor. Rod Block Monitor and Technical Snecification Imorovement (ARTS) Proaram for the Duane Arnold Enerav Center, NEDC-30813-P, December 1984.
5.
Apolication of the "Reaional Exclusion with Flow-Biased APRM Neutron Flux Scram" Stability Solution (Ootion I-D) to the Duane Arnold Enerav Center, GENE-A00-04021-01, September 1995.
- Latest NRC-approved revision.
I Amendment No. 120,142,180,183, 3.3-15 215
DAEC-1 (2)
R2sults of the last isotopic analysis for radiciodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than limit.
Each result should include date and time of sampling and radiciodine concentrations; (3)
Cleanup system operating status starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4)
Graph of I-131 concentration and one other radiciodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5)
The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.
6.11.2 CORE OPERATING LIMITS REPORT a.
Core cycle-dependent limits shall be established prior to each l
reload cycle, or prior to any remaining part of a reload cycle, for the following:
1 1)
Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)
- Specification 3.12.A.
1 2)
Linear Heat Generation Rate (LHGR) - Specification 3.12.B.
3)
Minimum Critical Power Ratio (MCPR) - Specification 3.12.C.
4)
MAPFAC, AND MAPFAC, tion 3.3.F.4.a. Factors which multiply the MAP j
limits - Specifica
}
5)
Exclusion Region in the power / flow map - Specification 3.3.F.3.
These limits shall be documented in the CORE OPERATING LIMITS REPORT.
b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in General Electric Standard Application for Reactor Fuel, NEDE-240ll-P-A (GESTAR II).*
l c.
The core operating limits shall be determined such that all applicable limits (a fuel thermal-mechanical limits, core thermal hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
- Approved revision number at time reload fuel analyses are performed.
i Amendment No. 100,120,157,170, 6.11-4 104,195, 215 i